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Radiation Protection & Shielding
The Radiation Protection and Shielding Division is developing and promoting radiation protection and shielding aspects of nuclear science and technology — including interaction of nuclear radiation with materials and biological systems, instruments and techniques for the measurement of nuclear radiation fields, and radiation shield design and evaluation.
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ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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The Standards Committee is responsible for the development and maintenance of voluntary consensus standards that address the design, analysis, and operation of components, systems, and facilities related to the application of nuclear science and technology. Find out What’s New, check out the Standards Store, or Get Involved today!
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Latest News
First astatine-labeled compound shipped in the U.S.
The Department of Energy’s National Isotope Development Center (NIDC) on March 31 announced the successful long-distance shipment in the United States of a biologically active compound labeled with the medical radioisotope astatine-211 (At-211). Because previous shipments have included only the “bare” isotope, the NIDC has described the development as “unleashing medical innovation.”
Ryan Kelly, Pavel Tsvetkov, Sunil Chirayath, John Poston, Evans Kitcher
Nuclear Technology | Volume 190 | Number 1 | April 2015 | Pages 72-87
Technical Paper | Radiation Transport and Protection | doi.org/10.13182/NT13-155
Articles are hosted by Taylor and Francis Online.
The objective of this paper is to analyze the uncertainty in radiation dose rate estimates outside of a used nuclear fuel (UNF) dry cask storage unit due to the parametric variability of concrete compositions and densities. This requires the selection of a limited number of concrete compositions from a standardized database and the development of a reference dry cask model, which can be used to estimate dose rate from neutrons and gamma rays. The model was developed for use in calculations with MCNP, with reference data from a UNF assembly source and geometry details based on the Holtec HI-STORM 100S UNF dry cask storage system, both provided by the Comanche Peak nuclear power plant. A number of cases have been developed and analyzed to compare the effects of variations in concrete compositions considering nominal densities, variations from nominal densities, fixed densities regardless of the specific composition, and variations in the decay source. The analysis confirmed that the parametric variability of concrete compositions is a major source of uncertainty in evaluations of dry cask dose rates. While precise results depend on the compositions compared, general trends can be identified. The largest fraction of the dose value in all cases, typically 70%, is due to gamma rays produced by the fission products. Density variation had a dominant effect on the dose rate. Composition variations, while density was held fixed, indicated that the specific composition data significantly impact the dose rates produced by neutrons and associated capture gamma rays. The impact due to composition variation on neutron dose rate was found to be on the order of 70% or higher for these test cases. The analysis indicates that uncertainties in concrete characteristics at the time of on-site pour procedures impact the actual shielding efficiency and, therefore, must be evaluated.
